All the structural components being used in nuclear power plants that are into contact with coolant water are made from austenitic Fe—Cr—Ni alloys. For example, coolant pipes, core barrels, baffle former bolts are Fe-based austenitic Fe—Cr—Ni alloys such as, for example, 300 series austenitic stainless steels. All these structural components that are into contact with coolant are made from austenitic Fe—Cr—Ni alloys because of their superior resistance to intergranular stress corrosion cracking (IGSCC).
However, when the age of the nuclear power plants exceeds 20 years, austenitic Fe—Cr—Ni alloys turn out to be susceptible particularly to IGSCC in which cracking occurs mainly along grain boundaries. Intergranular (IG) cracking of the structural components in primary water reactor environment is particularly termed ‘primary water stress corrosion cracking (PWSCC)’ or ‘intergranular stress corrosion cracking’, while IG cracking in neutron irradiation environment is called ‘irradiation assisted stress corrosion cracking (IGSCC)’. The above-mentioned cracking phenomena are often collectively referred to as ‘Intergranular (IG) cracking’. As the lifetime of the nuclear power plants is being extended from 40 to 60 years or even beyond, the aging management of nuclear power plants is particularly one of the hot issues to assure the safety of the aged nuclear power plants. From the perspective of the aging management of the nuclear power plants, the core technology particularly relates to the management of IG cracking of the structural components made of austenitic Fe—Cr—Ni alloys which are exposed to primary water.
Unfortunately, no particular technology has been made available so far, in association with the management of IG cracking, albeit either active or proactive, mainly because the intergranular cracking mechanism of austenitic Fe—Cr—Ni alloys has not been fully clarified yet. Instead, the nuclear industry focuses on repair technology of degradated structural components by overlay welding, rather than the proactive management of IG cracking of the structural components itself. Although efforts are made to develop alternative techniques to detect and monitor intergranular cracks in the austenitic Fe—Cr—Ni alloy structural components before propagation of IG cracks into a through-wall crack, IG cracks are too fine to be detected with high reliability by the conventional non-destructive examination techniques. In other words, non-destructive examination techniques to detect and monitor IG cracks of the structural components made of austenitic Fe—Cr—Ni alloys is the core technology for the aging management of nuclear power plants, but no current techniques available so far has fully resolved technical difficulties related to non-destructive examination of IG cracks despite much efforts put forward to date.
Recently, the present inventors showed that austenitic 316L stainless steel, one of the structural materials being used for the structural components of nuclear power plants, had lattice contraction to some extents in reactor operating conditions, due to short range ordering (Young Suk Kim et al., Transactions of the Korean Nuclear Society Autumn Meeting, Jeju Korea, 2010, pp. 1079-1080.)
Accordingly, given the inventors' observation, the inventors proposes a hypothesis that short range order occurring in austenitic Fe—Cr—Ni alloys during plant operation would cause intergranular stress corrosion cracking, and that intergranular stress corrosion cracking susceptibility of the structural components of metal alloys for nuclear power plants can be evaluated non-destructively by measuring variations in their properties accompanying SRO. In short, the inventors attempts to establish the non-destructive evaluation methods to detect and monitor intergranular stress corrosion cracks in the structural components made of austenitic Fe—Cr—Ni alloys and furthermore to assess the remaining lifetime of the structural components made of austenitic Fe—Cr—Ni alloys by accounting for the variations in their properties.